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Event Notification Report for May 26, 2000

                    U.S. Nuclear Regulatory Commission
                              Operations Center

                              Event Reports For
                           05/25/2000 - 05/26/2000

                              ** EVENT NUMBERS **

36888  37031  37032  37033  37034  

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|Power Reactor                                    |Event Number:   36888       |
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| FACILITY: WATERFORD                REGION:  4  |NOTIFICATION DATE: 04/13/2000|
|    UNIT:  [3] [] []                 STATE:  LA |NOTIFICATION TIME: 17:13[EDT]|
|   RXTYPE: [3] CE                               |EVENT DATE:        04/13/2000|
+------------------------------------------------+EVENT TIME:        15:35[CDT]|
| NRC NOTIFIED BY:  E. LEMKE                     |LAST UPDATE DATE:  05/25/2000|
|  HQ OPS OFFICER:  JOHN MacKINNON               +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |CHARLES PAULK        R4      |
|10 CFR SECTION:                                 |                             |
|AINC 50.72(b)(2)(iii)(C) POT UNCNTRL RAD REL    |                             |
|AOUT 50.72(b)(1)(ii)(B)  OUTSIDE DESIGN BASIS   |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
|                                                   |                          |
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                                   EVENT TEXT                                   
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| FEEDWATER ISOLATION VALVES MAY CLOSE FASTER THAN 1.5-SECOND DESIGN BASIS     |
| LIMIT.                                                                       |
|                                                                              |
| On March 21, 2000, Waterford Unit 3 determined that, based on a new          |
| calculation methodology and the latest stroke time data, the Feedwater       |
| Isolation Valves (FWIVs) FW-184 A (B) may close faster than the 1.5-second   |
| design basis limit.                                                          |
|                                                                              |
| The physical plant was determined to be operable and the FWIVs would have    |
| performed their intended safety function at the time the condition was       |
| identified.  An initial operability evaluation was made in accordance with   |
| procedure W4.101, which determined that the valves were operable at the time |
| the evaluation was conducted.  This was based on an engineering evaluation   |
| that determined that the faster closure of the FWIVs would not result in     |
| water hammer loads that would prevent the FWIVs and their associated         |
| penetrations from performing their required safety function.  The            |
| engineering evaluation determined that the analyzed increase in fast valve   |
| closure (FVC) load is 57% for FW-184A and 52.9% for FW-184B.                 |
|                                                                              |
| A subsequent evaluation was performed for FW-184A (B) to determine if at any |
| time in the last two years the increase in FVC load may have exceeded the    |
| allowable loads determined by the engineering evaluation for W4.101.  That   |
| evaluation determined on one occasion for FW-184A and six occasions for      |
| FW-184B, the percent increases for the FWIVs and subsequent stroke times     |
| exceeded the values provided in the operability determination                |
|                                                                              |
| On these occasions in question, closure of FW-184A (B) in response to the    |
| most adverse accident scenario could have potentially produced water hammer  |
| that may have exceeded the capability of  piping supports in the FW system   |
| between the SGs and FW-184A (B). This may have resulted in the subsequent    |
| loss of containment isolation function of FW-184A (B).                       |
|                                                                              |
| The NRC Resident Inspector was notified of this event by the licensee.       |
|                                                                              |
| * * * UPDATE ON 05/25/00 AT 1630 ET BY CHRIS PICKERING TAKEN BY MACKINNON *  |
| * *                                                                          |
|                                                                              |
| Entergy is revising this event notification.  The event was initially        |
| reported under 10CFR50.72(b)(2)(iii) as a condition that alone could have    |
| prevented fulfillment of a safety function.  Based upon further review of    |
| the condition, Entergy has determined the condition is reportable under      |
| 10CFR50.72(b)(1)(ii)(B) as a condition that was outside the design basis of  |
| the plant.                                                                   |
|                                                                              |
| Evaluation of the event determined that the potential impact on containment  |
| integrity as a result of this condition does not affect the results of the   |
| currently analyzed MSLB event's radiological consequences.  Further, the     |
| possibility of containment integrity loss due to excessive waterhammer loads |
| during a FWLB remains bounded by the MSLB with respect to dose consequences. |
| The FWLB event radiological consequences are bounded by the MSLB event,      |
| which is predicted to be within 10CFR100 limits.                             |
|                                                                              |
| This event posed a containment integrity condition that was outside the      |
| design basis of the plant.  An evaluation determined that the FWIVs could    |
| have potentially closed faster than their design basis minimum limit on      |
| seven occasions, six instances  for FW-184A and one instance for FW-184B.    |
| The original event notification (above) incorrectly noted that there was one |
| instance for FW-184A and six instances for FW-184B.  The faster valve        |
| closure times could have placed increased loads on the piping and piping     |
| supports in the FW system between the steam generators (inside containment)  |
| and the FWIVs (outside containment).  These increased loads were not         |
| accounted for in the design basis of the system.  Based on this information, |
| the event is reportable under 10CFR50.72(b)(1)(ii)(B) as a condition that    |
| was outside the design basis of the plant.                                   |
|                                                                              |
| The NRC Resident Inspector was notified of the event revision by the         |
| licensee.                                                                    |
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|Power Reactor                                    |Event Number:   37031       |
+------------------------------------------------------------------------------+
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| FACILITY: CATAWBA                  REGION:  2  |NOTIFICATION DATE: 05/25/2000|
|    UNIT:  [1] [] []                 STATE:  SC |NOTIFICATION TIME: 09:21[EDT]|
|   RXTYPE: [1] W-4-LP,[2] W-4-LP                |EVENT DATE:        05/25/2000|
+------------------------------------------------+EVENT TIME:        08:32[EDT]|
| NRC NOTIFIED BY:  KEVIN PHILLIPS               |LAST UPDATE DATE:  05/25/2000|
|  HQ OPS OFFICER:  FANGIE JONES                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |MARK LESSER          R2      |
|10 CFR SECTION:                                 |                             |
|AOUT 50.72(b)(1)(ii)(B)  OUTSIDE DESIGN BASIS   |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
|                                                   |                          |
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                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| AUXILIARY BUILDING FILTERED EXHAUST MAY NOT HAVE BEEN ABLE TO PERFORM ITS    |
| DESIGN FUNCTION                                                              |
|                                                                              |
| The licensee discovered both doors on the vestibule to the '1A' Centrifugal  |
| Charging Pump Room not fully closed, this may have prevented the Auxiliary   |
| Building filtered exhaust system from maintaining a negative pressure in     |
| that room.  This would render both trains of Auxiliary Building exhaust      |
| inoperable per Technical Specifications 3.7.12 and would require entry into  |
| Technical Specification 3.0.3.  The doors were closed on discovery which     |
| returned the unit to compliance with Technical Specifications.               |
|                                                                              |
| The licensee will followup with an investigation into when and how the doors |
| came to be open.                                                             |
|                                                                              |
| The licensee notified the NRC Resident Inspector.                            |
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+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   37032       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: MCGUIRE                  REGION:  2  |NOTIFICATION DATE: 05/25/2000|
|    UNIT:  [1] [] []                 STATE:  NC |NOTIFICATION TIME: 21:47[EDT]|
|   RXTYPE: [1] W-4-LP,[2] W-4-LP                |EVENT DATE:        05/25/2000|
+------------------------------------------------+EVENT TIME:        20:46[EDT]|
| NRC NOTIFIED BY:  H. M. HARRIS                 |LAST UPDATE DATE:  05/25/2000|
|  HQ OPS OFFICER:  JOHN MacKINNON               +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |MARK LESSER          R2      |
|10 CFR SECTION:                                 |JOHN HANNON          NRR     |
|ARPS 50.72(b)(2)(ii)     RPS ACTUATION          |JOSEPH GIITTER       IRO     |
|AESF 50.72(b)(2)(ii)     ESF ACTUATION          |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     A/R        Y       100      Power Operation  |0        Hot Standby      |
|                                                   |                          |
|                                                   |                          |
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                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| TURBINE RUNBACK TO 50% DUE TO LOSS OF A MAIN FEEDWATER PUMP FOLLOWED BY A    |
| REACTOR TRIP ON LOW LOW STEAM GENERATOR WATER LEVEL.  ONE AUXILIARY          |
| FEEDWATER PUMP HAD TO BE MANUALLY STARTED.                                   |
|                                                                              |
| Reactor automatic trip occurred at 2046 ET.  Initiating event was steam      |
| generator low low water level on "C" steam generator.  Prior to the trip     |
| Channel 1 vital 120 volt AC power was lost due to the tripping open of  EVIA |
| (DC/AC ) inverter AC output breaker.  No testing was in progress at this     |
| time and the cause of the AC breaker trip is under investigation.  When the  |
| AC output breaker opened it caused a loss of Channel 1 power.   Main         |
| feedwater pump "1A" control circuitry interfaces with Channel 1 120 volt AC  |
| power and when Main Feedwater pump "1A" control circuitry power was lost     |
| main feedwater pump "1A" turbine tripped.  The loss of Main Feedwater  pump  |
| "1A" turbine initiated an automatic main turbine runback to 50% power.       |
| After the main turbine runback to 50% power the reactor tripped on steam     |
| generator "1C" low low water level. Only one of two motor driven auxiliary   |
| feedwater pumps automatically started on steam generator "1C" low low water  |
| level.  Auxiliary feedwater pump "1A" was manually started approximately 2   |
| minutes after the automatic reactor trip. The licensee is investigating why  |
| the "1A" auxiliary feedwater pump did not automatically start.  All rods     |
| fully inserted into the core and reactor coolant temperature is being        |
| maintained at Tave no load condition of 557 degrees F.  No PORVs or code     |
| safety valves on either the primary or secondary side of the plant opened.   |
| All the Emergency Core Cooling Systems and the Emergency Diesel Generators   |
| are fully operable if needed.  Offsite electrical grid is stable.    The     |
| licensee's investigation into the initiating event is continuing.            |
|                                                                              |
| The NRC Resident Inspector was notified of this event by the licensee.       |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Fuel Cycle Facility                              |Event Number:   37033       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT   |NOTIFICATION DATE: 05/25/2000|
|   RXTYPE: URANIUM ENRICHMENT FACILITY          |NOTIFICATION TIME: 23:52[EDT]|
| COMMENTS: 2 DEMOCRACY CENTER                   |EVENT DATE:        05/25/2000|
|           6903 ROCKLEDGE DRIVE                 |EVENT TIME:        18:00[EDT]|
|           BETHESDA, MD 20817    (301)564-3200  |LAST UPDATE DATE:  05/26/2000|
|    CITY:  PIKETON                  REGION:  3  +-----------------------------+
|  COUNTY:  PIKE                      STATE:  OH |PERSON          ORGANIZATION |
|LICENSE#:  GDP-2                 AGREEMENT:  N  |JOHN MADERA          R3      |
|  DOCKET:  0707002                              |WAYNE HODGES         NMSS    |
+------------------------------------------------+                             |
| NRC NOTIFIED BY:  JIM MCCLEERY                 |                             |
|  HQ OPS OFFICER:  FANGIE JONES                 |                             |
+------------------------------------------------+                             |
|EMERGENCY CLASS:          N/A                   |                             |
|10 CFR SECTION:                                 |                             |
|NBNL                     RESPONSE-BULLETIN      |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
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                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| NRC BULLETIN 91-01, 24 HOUR NOTIFICATION                                     |
|                                                                              |
| The following is the faxed report from Portsmouth:                           |
|                                                                              |
| On May 16, 2000, PORTS plant personnel initiated a review of an NRC event (# |
| 36993) submitted by Paducah Gaseous Diffusion Plant (PGDP), to assess        |
| applicability at the Portsmouth sight. Errors associated with calibration    |
| correction factors used to determine uranium masses were found to have not   |
| occurred at the PORTS sight.                                                 |
|                                                                              |
| On May 25. 2000, after further review by Nuclear Criticality Safety dept.    |
| personnel, a violation of double contingency was identified when it was      |
| determined that the independence of non-destructive assay (NDA) measurements |
| was not maintained. NCSA-PLANT013.A00 indicated that NDA measurements will   |
| be maintained independent. However, calibration of instruments was not       |
| maintained independent and therefore the resulting measurements were not     |
| independent. As a result, the calibration process failed to ensure that a    |
| single failure could not effect two independent mass measurements. Double    |
| contingency was violated in cases where two independent uranium mass         |
| measurements were required to establish double contingency.                  |
|                                                                              |
| A review of NDA calibration data confirmed that no measurement errors        |
| actually existed which would have affected the mass readings taken to        |
| establish double contingency. This event is being reported because the       |
| independence or mass measurement readings was not established and            |
| maintained.                                                                  |
|                                                                              |
| SAFETY SIGNIFICANCE OF EVENTS:                                               |
|                                                                              |
| The safety significance of this event is very low. All historical            |
| calibrations were reviewed (Ref. POEF-38.340.00.086) and no discrepancies    |
| were identified with historical measurements. Furthermore, ongoing           |
| laboratory Quality Assurance programs and practices help ensure errors of    |
| this nature remain sufficiently unlikely to be relied on as criticality      |
| controls.                                                                    |
|                                                                              |
|                                                                              |
| POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO[S] OF HOW            |
| CRITICALITY COULD OCCUR):                                                    |
|                                                                              |
| An error in NDA measurements plus uncertainty would have to be low by        |
| greater than a factor of two or results in a critical mass of uranium going  |
| undetected. if this mass were then subject to the correct geometry,          |
| moderation and reflection conditions a criticality could error               |
|                                                                              |
|                                                                              |
| CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):     |
|                                                                              |
| Mass is the only controlled parameter. Two independent measurements of mass  |
| are required to ensure double contingency. Since independence of NDA of NDA  |
| mass measurements is called Into question, only one independent estimate of  |
| mass is available. The absence of a second independent NDA mass measurement  |
| represents a loss of one double contingency control.                         |
|                                                                              |
|                                                                              |
| ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS     |
| LIMIT AND % WORST CASE OF CRITICAL MASS):                                    |
|                                                                              |
| No uranium was involved in the event. NDA measurements are used to classify  |
| cascade deposits as either greater than safe mass or less than safe mass.    |
| (Safe Mass is 43.5% of minimum critical mass.) NDA measurements are also     |
| used in batching operations to ensure the 235U mass in the final container   |
| is less than 350 grams 235U which is less half the minimum critical mass at  |
| 100% enrichment and optimum moderation, geometry and reflection conditions.  |
|                                                                              |
|                                                                              |
| NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION   |
| OF THE FAILURES OR DEFICIENCIES                                              |
|                                                                              |
| Nuclear criticality safety controls include independent mass measurements of |
| 235U. Failure to ensure the calibration of the NDA instrumentation was       |
| independent lead to a potential failure of the mass measurements.            |
|                                                                              |
|                                                                              |
| CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEM AND WHEN EACH WAS IMPLEMENTED:   |
|                                                                              |
| NDA measurements performed for criticality safety purposes have been         |
| suspended until independent verification of calibration data requirements is |
| flown into applicable procedures and programs.                               |
|                                                                              |
| The certificate holder notified the NRC Resident Inspector and will notify   |
| the DOE representative.                                                      |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   37034       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: HOPE CREEK               REGION:  1  |NOTIFICATION DATE: 05/26/2000|
|    UNIT:  [1] [] []                 STATE:  NJ |NOTIFICATION TIME: 02:29[EDT]|
|   RXTYPE: [1] GE-4                             |EVENT DATE:        05/25/2000|
+------------------------------------------------+EVENT TIME:        22:54[EDT]|
| NRC NOTIFIED BY:  CHRIS SERATA                 |LAST UPDATE DATE:  05/26/2000|
|  HQ OPS OFFICER:  FANGIE JONES                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |DAN HOLODY           R1      |
|10 CFR SECTION:                                 |                             |
|AIND 50.72(b)(2)(iii)(D) ACCIDENT MITIGATION    |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     N          Y       15       Power Operation  |15       Power Operation  |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| FOUR FILTRATION RECIRCULATION AND VENTILATION FANS TRIPPED UNEXPECTEDLY      |
|                                                                              |
| "On 5/25/00 at 22:54 hours, four operating Filtration, Recirculation. and    |
| Ventilation System (FRVS) recirculation fans (A, B, D, E) unexpectedly       |
| tripped while they were running for a monthly surveillance. An FRVS          |
| ventilation fan, that was also in service, continued to run.                 |
|                                                                              |
| "The FRVS consists of two subsystems, the Recirculation System and the       |
| Ventilation System. The FRVS Recirculation System is an Engineered Safety    |
| Feature (ESF) System, located inside the Secondary Containment, that reduces |
| offsite doses significantly below 10CFR100 guidelines during a LOCA,         |
| refueling accident, or high radioactivity in the Secondary Containment. Upon |
| a Secondary Containment isolation, the FRVS Recirculation System is actuated |
| and recirculates the Secondary Containment air through filters for cleanup.  |
| This subsystem is the initial cleanup system before discharge is made via    |
| the FRVS Ventilation subsystem to the outdoors. The FRVS Ventilation System  |
| is an ESF system, located inside the Secondary Containment, that maintains   |
| the building at a negative pressure with respect to the outdoors. The system |
| takes suction from the discharge duct of the FRVS Recirculation system and   |
| discharges the air through filters to the outdoors.                          |
|                                                                              |
| "Investigation into the cause of the fan trips has identified a manual       |
| damper in the ventilation system ductwork that failed to the closed          |
| position. This manual damper is normally open during power operation. This   |
| damper is repositioned closed during refueling outages to redistribute       |
| ventilation through the Secondary Containment. Immediate actions were taken  |
| to return the damper to the open position. Recirculation fans have been      |
| returned to service and are operating satisfactorily.                        |
|                                                                              |
| "The ventilation system ductwork was reviewed to identify other manual       |
| dampers that could have the same or similar affect on system operation.  Two |
| other dampers were identified, their positions have been verified to be      |
| correct.                                                                     |
|                                                                              |
| "Hope Creek is presently in Operational Condition 1 at 15% power with the    |
| turbine/generator off-line awaiting replacement of the 'C' Main Power        |
| Transformer.  All safety related equipment is available."                    |
|                                                                              |
| The fans had been operating about 10 hours before tripping, it is suspected  |
| that the damper was not adequately secured in the open position when the     |
| plant exited the refueling outage.  The licensee is investigating.           |
|                                                                              |
| The licensee notified the NRC Resident Inspector and the local township      |
| authorities.                                                                 |
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